muap-09005-np, rev. 1, 'summary of stress analysis results

Summary of Stress Analysis Results for the US-APWR Reactor Vessel MUAP-09005-NP (R1)

Mitsubishi Heavy Industries, LTD.

Summary of Stress Analysis Results for the US-APWR Reactor Vessel

Non-proprietary Version

February 2011

Ⓒ2011 Mitsubishi Heavy Industries, Ltd.

All Rights Reserved

Summary of Stress Analysis Results for . the US-APWR Reactor Vessel MUAP~09005-NP (R1)

. Prepared: k.(7~ 2-/~/711

Kunimitsu Tatsuno, Engineering Manager Date . Component Designing Section Nuclear Plant Component Designing Department

Reviewed: . /J-r;t-f f~ r-~~ <f,:bll

Takatoshi Hirota, Acting Manager Date Component Designing Section Nuclear Plant Component Designing Department

Approved: . ~,~, ,r

.J- 'f-.{ J

Mastiyuki Mukai, Manager Date Component Designing Section Nuclear Plant Component Designing Department

Approved: :zl.~ ... ~-k. Z' if Hidehito Mimaki, Deputy Manager Date Nuclear Plant Component Designing Department

Approved: I:i /~ ~'. Fe-b.. ?l ~II Hitoshi Kaguchi:Mfn~ Date Nuclear Plant Component Designing D~partment

Mitsubishi Heavy Industries, L TO.



Revision History

Revision Page Description

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Original Issue

1

Abstract

Table of Contents

List of Tables

List of Figures

1.0 Introduction

2.0 Summary of results

3.0 Conclusions

7.2 Material

7.3 Loads, Load Combinations, and Transients

8.2 Stress Analysis

9.0 Computer Programs Used

10.0 Structual Analysis Results Abstract

12.0 References

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Delete analysis results of DVI nozzle and Vent Pipe, and revise analysis results.

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Arrange tables for material property.

Delete loads of DVI nozzle and Vent Pipe, and revise loads.

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© 2011

MITSUBISHI HEAVY INDUSTRIES, LTD.

All Rights Reserved

This document has been prepared by Mitsubishi Heavy Industries, Ltd. (“MHI”) in connection with the U.S. Nuclear Regulatory Commission’s (“NRC”) licensing review of MHI’s US-APWR nuclear power plant design. No right to disclose, use or copy any of the information in this document, other than by the NRC and its contractors in support of the licensing review of the US-APWR, is authorized without the express written permission of MHI.

This document contains technology information and intellectual property relating to the US-APWR, and it is delivered to the NRC on the express condition that it not be disclosed, copied or reproduced in whole or in part, or used for the benefit of anyone other than MHI without the express written permission of MHI, except as set forth in the previous paragraph.

This document is protected by the laws of Japan, U.S. copyright law, international treaties and conventions, and the applicable laws of any country where it is being used.

MITSUBISHI HEAVY INDUSTRIES, LTD. 16-5, Konan 2-chome, Minato-ku

Tokyo 108-8215 Japan



Abstract

This report contains a summary of the structural evaluation of the three Reactor Vessel (RV) parts.

The evaluations performed were based on the loading conditions defined in the US-APWR Reactor Vessel Design Specification (Reference 4) and on the design procedures per the 2001 Edition of Section III of the ASME Boiler & Pressure Vessel Code up to and including the 2003 addenda, referred to as the ASME Code of Reference, Section III (Reference 1).

The analyses and results presented demonstrate that the three parts of the RV that were evaluated satisfy all of the applicable structural limits of the ASME Code of Reference.


Mitsubishi Heavy Industries, LTD. i

Table of Contents

1.0 INTRODUCTION ..................................................................................................1-1

2.0 SUMMARY OF RESULTS....................................................................................2-1

3.0 CONCLUSIONS ...................................................................................................3-1

4.0 NOMENCLATURE................................................................................................4-1

5.0 ASSUMPTIONS AND OPEN ITEMS....................................................................5-1

5.1 ASSUMPTIONS...............................................................................................5-1

5.2 OPEN ITEMS..................................................................................................5-1

6.0 ACCEPTANCE CRITERIA....................................................................................6-1

7.0 DESIGN INPUT ....................................................................................................7-1

7.1 GEOMETRY ...................................................................................................7-1

7.2 MATERIAL .....................................................................................................7-1

7.3 LOADS, LOAD COMBINATIONS, AND TRANSIENTS ............................................7-5

7.3.1 Pressure Loads and Temperature ...................................................7-5

7.3.2 Mechanical Loads............................................................................7-5

7.3.3 Thermal and Pressure Transient Loads ........................................7-17

7.3.4 Load Combinations........................................................................7-19

8.0 METHODOLOGY .................................................................................................8-1

8.1 HEAT TRANSFER COEFFICIENTS AND THERMAL ANALYSIS ..............................8-1

8.2 STRESS ANALYSIS.........................................................................................8-1

8.3 FATIGUE ANALYSIS MODEL AND METHOD.......................................................8-3

9.0 COMPUTER PROGRAMS USED ........................................................................9-1

10.0 STRUCTURAL ANALYSIS RESULTS................................................................10-1

10.1 INLET & OUTLET NOZZLES...........................................................................10-1

10.1.1 Modeling and Analysis...................................................................10-1


Mitsubishi Heavy Industries, LTD. ii

10.1.2 Stress Results................................................................................10-6

10.2 RVCH NOZZLE J-GROOVE WELD ...............................................................10-8

10.2.1 Modeling and Analysis.................................................................10-10

10.2.2 Stress Results..............................................................................10-13

10.3 REACTOR VESSEL FLANGE & STUD BOLTS.................................................10-14

10.3.1 Modeling and Analysis.................................................................10-17

10.3.2 Stress Results..............................................................................10-21

11.0 FRACTURE MECHANICS ASSESSMENT ........................................................11-1

12.0 REFERENCES ...................................................................................................12-1


Mitsubishi Heavy Industries, LTD. iii

List of Tables

Table 2-1 Summary of Most Limiting Results.........................................................................2-1

Table 6-1 Class 1 Component Stress Limits (other than Bolts) .............................................6-2

Table 6-2 Class 1 Bolt Stress Limits ......................................................................................6-3

Table 7-1 US-APWR RV Basic Design Drawing List .............................................................7-1

Table 7-2 Materials of Construction .......................................................................................7-1

Table 7-3 Material Properties for SA-508 Grade 3 Class 1....................................................7-2

Table 7-4 Material Properties for SA-533 Type B Class 1 .....................................................7-2

Table 7-5 Material Properties for SA-540 Grade B24 Class 3 (Stud Bolts & Washers) .........7-3

Table 7-6 Material Properties for SB-167 UNS N06690.........................................................7-3

Table 7-7 Material Properties for SA-182 Grade F316...........................................................7-4

Table 7-8 Material Properties for SB-167 UNS N06690 (Code Case N-698) ........................7-4

Table 7-9 Pressures and Temperatures.................................................................................7-5

Table 7-10-1 Loads Applied to Inlet Nozzle ...........................................................................7-6

Table 7-10-2 Loads Applied to Inlet Nozzle ...........................................................................7-7

Table 7-11-1 Loads Applied to Inlet Nozzle RV Support........................................................7-8

Table 7-11-2 Loads Applied to Inlet Nozzle RV Support........................................................7-9

Table 7-12-1 Loads Applied to Outlet Nozzle.......................................................................7-10

Table 7-12-2 Loads Applied to Outlet Nozzle.......................................................................7-11

Table 7-13-1 Loads Applied to Outlet Nozzle RV Support ...................................................7-12

Table 7-13-2 Loads Applied to Outlet Nozzle RV Support ...................................................7-13

Table 7-14 Loads Applied to RVCH Nozzle .........................................................................7-14

Table 7-15 Loads applied to Core Barrel Flange .................................................................7-15

Table 7-16 Loads applied to Upper Support Flange ............................................................7-15

Table 7-17 Loads applied to IHP Lug...................................................................................7-16


Mitsubishi Heavy Industries, LTD. iv

Table 7-18 Reactor Vessel Design Transients (1/2) ............................................................7-17

Table 7-18 Reactor Vessel Design Transients (2/2) ............................................................7-18

Table 7-19 Load Combinations ............................................................................................7-20

Table 9-1 Computer Program Description..............................................................................9-1

Table 10-1 Summary of Results for Inlet & Outlet Nozzles (1/2)..........................................10-6

Table 10-1 Summary of Results for Inlet & Outlet Nozzles (2/2)..........................................10-7

Table 10-2 Summary of Results for RVCH Nozzle J-Groove Welds..................................10-13

Table 10-3-1 Summary of Results for Reactor Vessel Flanges .........................................10-21

Table 10-3-2 Summary of Results for Stud Bolts ...............................................................10-22


Mitsubishi Heavy Industries, LTD. v

List of Figures

Figure 1-1 General Configuration of the US-APWR Reactor Vessel .....................................1-2

Figure 8-1 Stress Evaluation Process....................................................................................8-2

Figure 10-1-1 Inlet / Outlet Nozzles Arrangement................................................................10-2

Figure 10-1-2 Inlet & Outlet Nozzles Dimensions ................................................................10-3

Figure 10-1-3 Inlet & Outlet Nozzles RV Support Pads Dimensions....................................10-4

Figure 10-1-4 Inlet /Outlet Nozzle and DVI Nozzle Finite Element Model............................10-5

Figure 10-2-1 RVCH Nozzle Locations ................................................................................10-8

Figure 10-2-2 RVCH Nozzle J-Groove Weld Dimensions....................................................10-9

Figure 10-2-3 RVCH Nozzle Finite Element Model............................................................10-11

Figure 10-2-4 Finite Element Mesh for RVCH Nozzle J-groove Welds..............................10-12

Figure 10-3-1 Reactor Vessel Dimensions.........................................................................10-15

Figure 10-3-2 Reactor Vessel Flange & Stud Bolts Dimensions........................................10-16

Figure 10-3-3 Reactor Vessel Flange & Stud Bolt Finite Element Model...........................10-18

Figure 10-3-4 Reactor Vessel Flange & Stud Bolt Finite Element Model...........................10-19

Figure 10-3-5 Finite Element Model for RV Flange & Stud Bolt ........................................10-20


Mitsubishi Heavy Industries, LTD. vi

List of Acronyms

The following list defines the acronyms used in this document.

CRDM Control Rod Drive Mechanism

DCD Design Control Document

DVI Direct Vessel Injection

FEA Finite Element Analysis

FSRF Fatigue Strength Reduction Factor

ICIS Incore Instrumentation System

IHP Integrated Head Package

RCS Reactor Coolant System

RV Reactor Vessel

RVCH Reactor Vessel Closure Head

SRSS Square Root of Sum of Square

SSE Safe Shutdown Earthquake

TC Thermocouple


Mitsubishi Heavy Industries, LTD. 1-1

1.0 INTRODUCTION

This Stress Analysis Report contains a summary of the stress analysis results for the US-APWR Reactor Vessel (RV). The content of this report follows the ASME guidelines for Design Reports (Section III Division 1 Appendix C).

The RV is a vertical cylindrical vessel with hemispherical top and bottom heads. The top head or reactor vessel closure head (RVCH) is a removable, flanged closure head connected to the RV upper shell flange by stud bolts. The RV has four inlet nozzles, four outlet nozzles, and four direct vessel injection (DVI) nozzles.

The RVCH consists of a hemispherical dome welded to the closure head flange, which is a thick ring forging. The seal between the RVCH and RV upper shell flange is made by two metallic O-rings that are fitted into grooves machined in the RVCH flange cladding. The RVCH includes the control rod drive mechanism (CRDM) nozzles, the incore instrumentation system (ICIS) nozzles, the thermocouple (TC) nozzles and the vent pipe, which are welded to RVCH by partial penetration J-groove welds.

Figure 1-1 shows the general configuration of the US-APWR Reactor Vessel.

This report provides structural evaluations for three RV parts. The three evaluated RV parts are listed in Table 2-1. This Stress Analysis Report summarizes the stress results based upon detailed analyses that demonstrate that the RV components meet the requirements of the Design Specification (Reference 4).



Figure 1-1 General Configuration of the US-APWR Reactor Vessel

1. Inlet & Outlet Nozzles

2. RVCH Nozzle J-groove Welds 3. Reactor Vessel

Flange & Stud Bolts



2.0 SUMMARY OF RESULTS

The three evaluated parts of the RV, along with the most limiting results in each, are listed in Table 2-1 below. The structural analysis results for each of these parts are provided in Section 10.

Table 2-1 Summary of Most Limiting Results

Section Evaluated Part Max Stress /

Allowable Ratio1

Highest Fatigue2

Usage Factor

10.1 Inlet & Outlet Nozzles

10.2 RVCH Nozzle J-groove Weld

10.3 Reactor Vessel Flange & Stud Bolts

Note-1: The allowable ratio is the “ratio” of the calculated stress intensity to the allowable stress intensity. Therefore, any ratio smaller than or equal to 1.0 is acceptable.

IntensityStressAllowable

IntensityStress CalculatedRatio

Note-2: The fatigue calculations performed in this report meet the requirements of the ASME code. Environmental fatigue analysis as described in NCR RG 1.207, “Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components Due to the Effects of the Light-Water Reactor Environment for New Reactors” will be evaluated separately.



3.0 CONCLUSIONS

The US-APWR RV is designed to the requirements of the ASME Boiler and Pressure Vessel Code, 2001 Edition up to and including the 2003 Addenda for the Design, Service Loadings, Operating Conditions, and Test Conditions as specified in the Design Specification (Reference 4).



4.0 NOMENCLATURE

Symbol Unit Definition

Pm ksi General Primary Membrane Stress

PL ksi Local Primary Membrane Stress

Pb ksi Primary Bending Stress

Q ksi Secondary Stress

Sm ksi Design Stress Intensity

Sy ksi Yield Stress

Su ksi Tensile Strength

Ab in2 Actual Total Cross-Sectional Area of Bolts at Root of Thread or Section of Least Diameter Under Stress

Am in2 Required Total Design Cross-Sectional Area of Bolts, taken as the greater of Am1 and Am2

Am1 in2 Total cross-sectional area of bolts at root of thread or section of least diameter under stress, required for Design Condition =Wm1/Sb

Am2 in2 Total cross-sectional area of bolts at root of thread or section of least diameter under stress, required for gasket seating =Wm2/Sa

St ksi Averaged Stress for Bolt (neglecting stress concentration)

St + Sb ksi Tension plus Bending Stress for Bolt (neglecting stress concentration)

Wm1 lb Minimum required bolt load for Design Pressure

Wm2 lb Minimum required bolt load for gasket seating

yA - Thermal Ratcheting Factor

SS ksi Thermal Stress Range

- Shape Factor

P psi Design Pressure

1/3 SSE - Level B Service Loading Earthquake

SSE - Safe Shutdown Earthquake



5.0 ASSUMPTIONS AND OPEN ITEMS

5.1 Assumptions

The basic modeling assumptions from the detailed analyses are as follows:

1. The inside diameter is taken as the drawing nominal value.

2. The wall thickness is the drawing nominal value.

3. For particular cases, cladding less than 10% of the total thickness is not considered in the stress analysis.

5.2 Open Items

There are no open items in this Technical Report.



6.0 ACCEPTANCE CRITERIA

The stress intensity acceptance criteria for Class 1 components are specified in NB-3220, NB-3230, and Appendix F of ASME Code Section III. Table 6-1 lists the stress limits for other than bolts, and Table 6-2 lists the stress limits for bolts.



Table 6-1 Class 1 Component Stress Limits (other than Bolts)

Condition Stress Category Stress Limits Remarks

Pm Sm NB-3221.1 PL 1.5 Sm NB-3221.2

PL + Pb Sm1) 2) or 1.5 Sm NB-3221.3

Bearing Stress Sy 6) or 1.5 Sy

6) NB-3227.1(a) Shear Stress 0.6 Sm NB-3227.1(b)

Pure Shear Stress 0.6 Sm NB-3227.2(a)

Design

Triaxial Sress4) 4 Sm NB-3227.4 PL + Pb + Q 3 Sm

NB-3222.2 Thermal Ratchet, SS 5) Sy × yA NB-3222.5

Usage Factor 1.0 NB-3222.4 Bearing Stress Sy

6) or 1.5 Sy 6) NB-3227.1(a)

Shear Stress 0.6 Sm NB-3227.1(b) Pure Shear Stress 0.6 Sm NB-3227.2(a)

Level A & B

Triaxial Sress4) 4 Sm NB-3224.3 Pm 1.1 Sm NB-3223 PL 1.5 (1.1 Sm) NB-3223 Level B

PL + Pb (1.1Sm)1) 2) or 1.5 (1.1 Sm) NB-3223

Pm Max (1.2 Sm, Sy)

Max (1.1 Sm, 0.9 Sy)3)

NB-3224.1

PL Max (1.8 Sm, 1.5 Sy) NB-3224.1

PL + Pb Max ( (1.2 Sm), Sy)

1) 2)

or Max (1.8 Sm, 1.5 Sy) NB-3224.1




Level C

Triaxial Sress4) 4.8 Sm NB-3224.3

Pm For ferritic materials, 0.7 Su

For austenitic and high alloy steels, Min (2.4 Sm, 0.7 Su)

PL For ferritic materials, 1.5 (0.7 Su)

For austenitic and high alloy steels, 1.5 Min (2.4 Sm, 0.7 Su)

PL + Pb For ferritic materials, 1.5 (0.7 Su)

For austenitic and high alloy steels, 1.5 Min (2.4 Sm, 0.7 Su)

Level D

Pure Shear 0.42 Su

NB-3225 (Appendix F-1331.1)

Pm 0.9 Sy

Pm + Pb (1.35 Sy) - for Pm 0.67 Sy

(or 0.9 Sy for non-rectangular sections) (2.15 Sy -1.2 Pm) - for 0.67 Sy Pm 0.9 Sy

NB-3226




Test

Triaxial Sress4) 4 Sm NB-3227.4



Note-1 The shape factor of for solid rectangular sections is 1.5, shall not exceed 1.5. Note-2 “” is considered where stresses are classified as primary bending. Note-3 The stress limits for pressure loading alone for ferritic material. Note-4 NB-3227.4 states that the Triaxial Stress limit is 4 Sm and does not apply to Level D.

NB-3224.3 states the Level C limit is 4.8 Sm. Note-5 NB-3222.5 requires evaluation of Thermal Stress Ratcheting for Level A Service Loads. In all

cases where elastic analysis indicates that the primary membrane stress is less than Sm and the primary plus secondary stress is less than 3 Sm, then thermal stress ratcheting will not occur.

Note-6 Sy when the distance to a free edge is less than the distance over which the bearing load is applied; 1.5 Sy when the distance to a free edge is larger.

Table 6-2 Class 1 Bolt Stress Limits

Condition Stress Category Stress Limits Remarks

Design Ab Am NB-3231, E-1000

Average Service Stress 1), St 2 Sm NB-3232.1

Max Service Stress 1), St + Sb 3 Sm NB-3232.2 Level A & B

Fatigue Usage Factor 2) 1.0 NB-3232.3

Average Service Stress 1), St 2 Sm NB-3234 Level C

Max Service Stress 1), St + Sb 3 Sm NB-3234

Average Tensile Stress 3), St Min (Sy, 0.7 Su) NB-3235 & F-1335.1

Max Tensile Stress 3), St + Sb Su NB-3235 & F-1335.1

Average bolt shear Min (0.6 Sy, 0.42 Su) F-1335.2

Combined tensile and shear 12222 vbFvftbFtf4) F-1335.3

Distance from bolt center to edge d [0.5 + 1.2 (fp / Su)] 5) F-1335.4(a)

Level D

Nominal bearing stress 2.1 Su F-1335.4(b)

Note-1 Includes preload, pressure, and differential thermal expansion, excludes stress concentrations.

Note-2 Includes a fatigue strength reduction factor of 4 for the threads. Note-3 Includes preload, pressure, differential thermal expansion, and prying action

produced by deformation of the connected parts, excludes stress concentrations. Note-4 ft=computed tensile stress, fv=computed shear stress, Ftb=allowable tensile stress at

operating temperature, Fvb=allowable shear stress at operating temperature. Note-5 d= nominal bolt diameter; fp = nominal bearing stress.



7.0 DESIGN INPUT

7.1 Geometry

The US-APWR RV basic drawings used to supply dimensions for the stress analysis are listed in Table 7-1. Figures describing the detailed geometry of the parts evaluated are provided in Section 10.

Table 7-1 US-APWR RV Basic Design Drawing List

No. Drawing Title Drawing Number

1 Reactor Vessel Outline and Design Drawings N0-F100A00

7.2 Material

The materials for the pressure boundary and internals are listed in Table 7-2.

Table 7-2 Materials of Construction

Part or Assembly Material

Reactor Vessel, Nozzles & RVCH SA-508 Grade 3 Class 1

RV Nozzle Safe Ends SA-182 Grade F316

Stud Bolts, Nuts and Washers SA-540 Grade B24 Class 3

RV IHP Support & Lifting Lugs SA-533 Type B Class 1

RVCH Nozzles SA-167 UNS N06690 (Code Case N-698)

Vent Pipe SB-167 UNS N06690

Seal Ledge SA-240 Type 304

The material strength properties used in the stress analyses are provided in Table 7-3 through 7-8. The strength properties were obtained from ASME Section II (Reference 2).



Table 7-3 Material Properties for SA-508 Grade 3 Class 1

Temperature, °F Sm (ksi) Sy (ksi) Su (ksi)

70 26.7 50.0 80.0

100 26.7 50.0 80.0

200 26.7 47.0 80.0

300 26.7 45.5 80.0

400 26.7 44.2 80.0

500 26.7 43.2 80.0

600 26.7 42.1 80.0

650 26.7 41.5 80.0

Table 7-4 Material Properties for SA-533 Type B Class 1


70

100

200

300

400

500

600

650

LATER



Table 7-5 Material Properties for SA-540 Grade B24 Class 3 (Stud Bolts & Washers)


70

100

200

300

400

500

600

700

Table 7-6 Material Properties for SB-167 UNS N06690


70

100

200

300

400

500

600

650

LATER

LATER



Table 7-7 Material Properties for SA-182 Grade F316


70 20.0 30.0 70.0

100 20.0 30.0 70.0

200 20.0 25.9 70.0

300 20.0 23.4 68.0

400 19.3 21.4 67.1

500 18.0 20.0 67.0

600 17.0 18.9 67.0

650 16.6 18.5 67.0

Table 7-8 Material Properties for SB-167 UNS N06690 (Code Case N-698)


100

200

300

400

500

600

700

LATER



7.3 Loads, Load Combinations, and Transients

The loads, load combinations and transients are defined in the Design Specification. The following is a summary of those used for the RV structural evaluations.

7.3.1 Pressure Loads and Temperature

Table 7-9 Pressures and Temperatures

Parameter Value

Design pressure 2500 psia (2485 psig)

Design temperature 650oF

Normal Operating Pressure 2250 psia (2235 psig)

Normal Operating Outlet Water Temperature 617.0oF

Normal Operating Inlet Water Temperature 550.6oF

Hydrostatic test pressure 3122 psia (3107 psig)

Minimum Hydrostatic test temperature 70oF

7.3.2 Mechanical Loads

The external loads are dead weight, thermal expansion loads, seismic loads, and accident loads. The external loads are considered at the nozzles and supports of the pressure boundary and obtained from the Design Specification (Reference 4).

Tables 7-10-1 to 7-17 provide the external and internal loads on the RV.

The bolt preload values for the studs was the minimum required bolt load for the design pressure and hydrostatic pressure calculated in accordance with Article E-1000 of the ASME code plus a 10% margin.



Table 7-10-1 Loads Applied to Inlet Nozzle

No. Loading Fx

(kips)

Fy (kips)

Fz (kips)

Mx (kips・in)

My (kips・in)

Mz (kips・in)

Loop-A,D 1

Dead Weight

Loop-B,C

Gr.1

Gr.2

Gr.3

Gr.4

Gr.5

Gr.6

Gr.7

Gr.8

Gr.9-1

Loop-A,D

Gr.9-2

Gr.1

Gr.2

Gr.3

Gr.4

Gr.5

Gr.6

Gr.7

Gr.8

Gr.9-1

2 Thermal

Loop-B,C

Gr.9-2



Table 7-10-2 Loads Applied to Inlet Nozzle

No. Loading Fx (kips)

Fy (kips)

Fz (kips)

Mx (kips・in)

My (kips・in)

Mz (kips・in)

3 Seismic (1/3 SSE)

4 Seismic (SSE)

5 Accident 1



Table 7-11-1 Loads Applied to Inlet Nozzle RV Support

No. Loading Fx

(kips)

Fy (kips)

Fz (kips)

Mx (kips・in)

My (kips・in)

Mz (kips・in)

Loop-A,D

Loop-B,C 1

Dead Weight

Dynamic load of tripped control

rods

Gr.1

Gr.2

Gr.3

Gr.4

Gr.5

Gr.6

Gr.7

Gr.8

Gr.9-1

Loop-A,D

Gr.9-2

Gr.1

Gr.2

Gr.3

Gr.4

Gr.5

Gr.6

Gr.7

Gr.8

Gr.9-1

2 Thermal

Loop-B,C

Gr.9-2



Table 7-11-2 Loads Applied to Inlet Nozzle RV Support


Fy (kips)

Fz (kips)

Mx (kips・in)

My (kips・in)

Mz (kips・in)

3 Seismic (1/3 SSE)

4 Seismic (SSE)

5 Accident



Table 7-12-1 Loads Applied to Outlet Nozzle

No. Loading Fx

(kips)

Fy (kips)

Fz (kips)

Mx (kips・in)

My (kips・in)

Mz (kips・in)

Loop-A,D 1

Dead Weight

Loop-B,C

Gr.1

Gr.2

Gr.3

Gr.4

Gr.5

Gr.6

Gr.7

Gr.8

Gr.9-1

Loop-A,D

Gr.9-2

Gr.1

Gr.2

Gr.3

Gr.4

Gr.5

Gr.6

Gr.7

Gr.8

Gr.9-1

2 Thermal

Loop-B,C

Gr.9-2



Table 7-12-2 Loads Applied to Outlet Nozzle


Fy (kips)

Fz (kips)

Mx (kips・in)

My (kips・in)

Mz (kips・in)

3 Seismic (1/3 SSE)

4 Seismic (SSE)

5 Accident 1



Table 7-13-1 Loads Applied to Outlet Nozzle RV Support

No. Loading Fx

(kips)

Fy (kips)

Fz (kips)

Mx (kips・in)

My (kips・in)

Mz (kips・in)

Loop-A,D

Loop-B,C 1

Dead Weight

Dynamic load of tripped control

rods

Gr.1

Gr.2

Gr.3

Gr.4

Gr.5

Gr.6

Gr.7

Gr.8

Gr.9-1

Loop-A,D

Gr.9-2

Gr.1

Gr.2

Gr.3

Gr.4

Gr.5

Gr.6

Gr.7

Gr.8

Gr.9-1

2 Thermal

Loop-B,C

Gr.9-2



Table 7-13-2 Loads Applied to Outlet Nozzle RV Support


Fy (kips)

Fz (kips)

Mx (kips・in)

My (kips・in)

Mz (kips・in)

3 Seismic (1/3 SSE)

4 Seismic (SSE)

5 Accident



Table 7-14 Loads Applied to RVCH Nozzle


Fy (kips)

Fz (kips)

Mx (kips·in)

My (kips·in)

Mz (kips·in)

1 Dead Weight

Location1

2 Seismic (1/3 SSE)

Location2

Location1

3 SRSS (SSE+ Accidentt

Location2



Table 7-15 Loads applied to Core Barrel Flange


Fy (kips)

Fz (kips)

Mx (kips·in)

My (kips·in)

Mz (kips·in)

Weight of core and core support structure

Core clamping load

1 Mechanical Weight

Dynamic load of tripped control rods

2 Seismic (1/3 SSE)

3 SRSS (SSE+ Accident)(1)

Note1):

Table 7-16 Loads applied to Upper Support Flange


Fy (kips)

Fz (kips)

Mx (kips·in)

My (kips·in)

Mz (kips·in)

1 Mechanical Weight

Spring Reaction

2 Seismic (1/3 SSE)

3 SRSS (SSE+ Accident)



Table 7-17 Loads applied to IHP Lug


Fy (kips)

Fz (kips)

Mx (kips·in)

My (kips·in)

Mz (kips·in)

1 Dead Weight

2 Seismic (1/3 SSE)

3 SRSS (SSE+Accident)



7.3.3 Thermal and Pressure Transient Loads

The design transients used in the structural evaluation are listed in Tables 7-18. These transients were determined based on a 60-year plant operating period and are classified into the ASME Level A, Level B, Level C, and Level D service conditions, and Test conditions, depending on the expected frequency of occurrence and severity of the event.

Table 7-18 Reactor Vessel Design Transients (1/2)

Mark Transient Occurrence Remark

I-a Plant heat-up (50oF/h) 120

I-b Plant cooldown (100oF/h) 120

Includes Transient for Loss of offsite power with natural circulation cooldown (10 times) and Safe shutdown (1 time)

I-c-1 Ramp load increase between 15% and 100% of full power (5% or full power per minute)

600

I-c-2 Ramp load increase between 50% and 100% of full power (5% or full power per minute)

19,200

I-d-1 Ramp load decrease between 15% and 100% of full power (5% or full power per minute)

600

I-d-2 Ramp load decrease between 50% and 100% of full power (5% or full power per minute)

19,200

I-e Step load increase of 10% of full power

600

I-f Step load decrease of 10% of full power

600

I-g Large step load decrease with turbine bypass

60

i) Steady state fluctuations

1 x 106 PP±50psi, Thot,Tcold,Tave±3.1oF I-h

Steady-state fluctuations and load regulation

ii) Load regulation

8 x 105

I-i Main feedwater cycling 2,100

I-j Refueling 60 Water is replaced in 10 minutes

I-k Ramp load increase between 0% and 15% of full power

600

I-l Ramp load decrease between 0% and 15% of full power

600

I-m RCP startup 3,000

I-n RCP shutdown 3,000

I-o Core lifetime extension 60

I-p Primary leakage test 120

I-q Turbine roll test 10



Table 7-18 Reactor Vessel Design Transients (2/2)

Mark Transient Occurrence Remark

II-a Loss of load 60

II-b Loss of offsite power 60

II-c Partial loss of reactor coolant flow 30

i) With no inadvertent cooldown

60

ii) With cooldown and no safety injection

30 II-d Reactor trip from full power

iii) With cooldown and safety injection

10

II-e Inadvertent RCS depressurization 30

II-f Control rod drop 30

II-g Inadvertent safeguards actuation 30

II-h Emergency feedwater cycling 700

II-i Cold over-pressure 30

II-j Excessive feedwater flow — Covered by Transient for reactor trip from full power ii)

II-k Loss of offsite power with natural circulation cooldown

— Covered by Transient for plant cooldown

II-l Partial loss of emergency feedwater 30 Use Figure for Transient of loss of offsite power

II-m Safe Shutdown — Covered by Transient for plant cooldown.

III-a Small loss of coolant accident 5 III-b Small steam line break 5 III-c Complete loss of flow 5 III-d Small feedwater line break 5 III-e SG tube rupture 5 IV-a Large loss of coolant accident 1 IV-b Large steam line break 1 IV-c RCP locked rotor 1 IV-d Control rod ejection 1 IV-e Large feedwater line break 1 V-a Primary-side hydrostatic test 10



7.3.4 Load Combinations

The loading conditions consist of various combinations of pressure, temperature and external loads. The load combinations considered in the analyses are listed in Table 7-19.



Table 7-19 Load Combinations

System Operating Condition

Service Stress Limit

Service Loading Combination

Design Design Design Pressure

Dead Weight Loads

Mechanical Loads

Thermal Loads(1)

Normal Level A Level A Thermal & Pressure Transients

Dead Weight Loads

Mechanical Loads

Thermal Loads(1)

Upset Level B Level B Maximum Pressure(2)

Level B Thermal & Pressure Transients

Dead Weight Loads

Mechanical Loads

Thermal Loads(1)

Seismic(1/3 SSE) Loads

Emergency Level C Level C Maximum Pressure

Level C Thermal & Pressure Transients(3)

Dead Weight Loads

Mechanical Loads

Thermal Loads(1)

Faulted Level D Level D Maximum Pressure

Dead Weight Loads

Mechanical Loads

Thermal Loads(1)

±SRSS(Seismic(SSE) Loads + Accident Loads(4))

Test Test Dead Weight Loads

Hydrostatic Test Pressure

Note-1: Applied to the nozzle only within the limits of reinforcement for the primary stress evaluation.

Note-2: Applied to the primary stress evaluation.

Note-3: Applied to the bolts instead of Maximum Pressure.

Note-4: If more than one Accident Loads, each are to be analyzed separately.



8.0 METHODOLOGY

The ABAQUS computer program was used to determine loads, temperature distributions, stresses, and deformations of the structure. ABAQUS is a general purpose finite element computer program used by MHI in the design and analysis of nuclear components. ABAQUS is available in the public domain and has been used by MHI for U.S. replacement steam generator and replacement RV closure head projects.

8.1 Heat Transfer Coefficients and Thermal Analysis

Heat transfer coefficients on the inner and outer surfaces of the component are required to define the temperature distributions during transients. Classical Handbook heat transfer equations (Reference 6) were used to calculate the heat transfer coefficients.

Finite element thermal analyses were performed for all Level A and Level B transients to define the time-dependent temperature distributions in the structure. The RCS fluid temperature versus time curves were applied to all wetted surfaces with appropriate heat transfer coefficients as discussed above. The outside surfaces under the vessel insulation were considered to be adiabatic.

8.2 Stress Analysis

Finite element stress analyses were performed for given loads and boundary conditions. The thermal loads obtained from the thermal solution were input to each node of the structural model. The calculation of NB-3200 stress intensities, stress classifications, and stress evaluations were performed using a set of MHI proprietary computer codes (CLASS2D, CLASS3D, EDITSTRS, EVALPRI, EVALSEFAV, RATCHET, ASMETEMP and EB3500). These codes are described in Section 9.

Figure 8-1 describes the stress evaluation process.

CLASS2D and CLASS3D were used to evaluate the stresses resulting from pressure, thermal loads, and externally applied forces and moments. EDITSTRS creates input files for the stress evaluation programs EVALPRI, EVALSEFAV, and RATCHET. EVALPRI and EVALSEFAV quantify the primary stress intensities, quantity the primary plus secondary stress ranges, and perform the fatigue evaluation. The RATCHET program was used for the thermal ratchet evaluation.

Detailed assumptions associated with the finite element model development and mesh refinement are documented in the detailed calculations. Finite element models were verified by hand calculations using handbook equations.



Figure 8-1 Stress Evaluation Process



8.3 Fatigue Analysis Model and Method

The fatigue analysis was performed following the standard rules of NB-3216.2 and NB-3222.4(e) of ASME Code Section III. These rules require calculation of the total stress, including the peak stress, to determine the allowable number of stress cycles for the specified Service Loadings at every point in the structure. In some cases, such as the root of the J-weld, a fatigue strength reduction factor (FSRF) was used where the peak stress cannot be accurately calculated. In these cases, the factor was applied to the linearized membrane plus bending stress.

The design transients for ASME Level A and B service conditions (Tables 7-18) were used in the evaluation of fatigue caused by cyclic loading. The effect of 300 cycles of a 1/3 SSE seismic event was also included in the evaluation of fatigue, and treated as a Level B service condition. The number of cycles assumed for the 1/3 SSE seismic event was based on a fatigue usage factor equivalent to that for a single SSE event of 20 cycles. 120 stud bolt tensioning cycles were used in the analysis.



9.0 COMPUTER PROGRAMS USED

Refer to Figure 8-1 for a visual description of the Stress Evaluation Process. Table 9-1 provides a brief description of each of the computer programs used.

Table 9-1 Computer Program Description

No. Program

Name Version Description

1 ABAQUS 6.7-1 ABAQUS is a general purpose finite element computer code that performs a wide range of linear and nonlinear engineering simulations

2 CLASS2D 4.0 CLASS2D is an MHI Code for classifying the stresses for axisymmetric models

3 CLASS3D 4.0 CLASS3D is an MHI Code for classifying the stresses for 3D solid models

4 EDITSTRS 4.0 EDITSTRS is an MHI Code that creates input files for the stress evaluation programs

5 EVALPRI 7.0 EVALPRI is an MHI Code that performs the primary stress evaluation

6 EVALSEFAV 7.0 EVALSEFAV is an MHI Code that performs the secondary stress and fatigue evaluation

7 RATCHET 8.0 RATCHET is an MHI Code that evaluates thermal stress ratcheting

8 ASMETEMP 1.0 ASMETEMP is an MHI Code that creates temperature files for the stress evaluation program

9 EB3500 3.0 EB3500 is an MHI Code that performs the general primary membrane stress evaluation.

All these computer programs were verified and validated and are maintained in compliance with the MHI quality assurance program. The computer programs were validated using one of the methods described below. Verification tests demonstrate the capability of the computer program to produce valid results for the test problems encompassing the range of permitted usage defined by the program documentation.

Hand calculations

Known solution for similar or standard problem

Acceptable experimental test results

Published analytical results

Results from other similar verified programs



10.0 STRUCTURAL ANALYSIS RESULTS

This Section summarizes the results of the analyses for the three parts of the RV. The stress calculations were performed using a combination of finite element analysis (FEA) and hand calculations. The general element types used for creation of the finite element model were the 20-node quadratic heat transfer brick and the 20-node quadratic brick from the ABAQUS element library, for heat transfer analysis and structural analysis, respectively.

The calculated results were conservative compared to the values that would be determined if more detailed calculations were performed. However, since the ASME Code allowable limits were satisfied in all cases, further analysis was not necessary.

10.1 Inlet & Outlet Nozzles

This section describes the stress evaluation results for the RV Inlet & Outlet Nozzles. Four inlet and four outlet nozzles are located between the RV flange and the top of the core. Figures 10-1-1 and 10-1-2 show the nozzle dimensions. The RV support pads are integral elements of the nozzle forgings. During normal operation, the RV expands in the radial and axial directions and the supports accommodate the expansion. The inlet and outlet support pad dimensions are shown in Figure 10-1-3.

10.1.1 Modeling and Analysis

Finite element thermal analyses were performed for Level A and B conditions with the inlet nozzle and RV internal fluid temperature set at Tcold and the outlet nozzle internal fluid temperature at Thot.



Figure 10-1-1 Inlet / Outlet Nozzles Arrangement

Dimensions: inch



Figure 10-1-2 Inlet & Outlet Nozzles Dimensions

Dimensions: inch

Inlet Nozzle

Outlet Nozzle



Figure 10-1-3 Inlet & Outlet Nozzles RV Support Pads Dimensions

Dimensions: inch

Inlet Nozzle

Outlet Nozzle



Figure 10-1-4 Inlet /Outlet Nozzle and DVI Nozzle Finite Element Model



10.1.2 Stress Results

The Stress-to-Allowable Ratios (calculated stress divided by allowable value) for the most limiting locations in the Inlet & Outlet Nozzles are summarized in Table 10-1.

Table 10-1 Summary of Results for Inlet & Outlet Nozzles (1/2)

Stress-to-Allowable Ratio

Primary Stress Primary + Secondary

Stress

Thermal Ratchet

FatigueCondition Part

Pm PL or PL+Pb

Triaxial Stress

PL+Pb+Q Usage Factor

Inlet Nozzle End

Inlet Nozzle Body

Inlet Nozzle-to-shell junction

Outlet Nozzle End

Outlet Nozzle Body Outlet Nozzle-to-shell

junction

Inlet RV Support Pad-to-shell junction

Design

Outlet RV Support Pad-to-shell junction

Inlet Nozzle End

Inlet Nozzle Body


Outlet Nozzle End


junction


Level A / B


Inlet Nozzle End

Inlet Nozzle Body


Outlet Nozzle End


junction


Level C




Table 10-1 Summary of Results for Inlet & Outlet Nozzles (2/2)


Primary Stress Primary + Secondary

Stress

Thermal Ratchet

FatigueCondition Part

Pm PL or PL+Pb

Triaxial Stress


Inlet Nozzle End

Inlet Nozzle Body


Outlet Nozzle End


junction


Level D


Inlet Nozzle End

Inlet Nozzle Body


Outlet Nozzle End


junction


Test




10.2 RVCH Nozzle J-Groove Weld

Figure 10-2-1 RVCH Nozzle Locations



Figure 10-2-2 RVCH Nozzle J-Groove Weld Dimensions



Figure 10-2-3 RVCH Nozzle Finite Element Model



Figure 10-2-4 Finite Element Mesh for RVCH Nozzle J-groove Welds




The Stress-to-Allowable Ratio (calculated stress divided by the allowable value) for the most limiting locations in the RVCH Nozzle J-groove Welds are summarized in Table 10-2.

Table 10-2 Summary of Results for RVCH Nozzle J-Groove Welds


Primary Stress Primary plus Secondary

Stress Condition

Pm PL or

PL+Pb Triaxial Stress

PL+Pb+Q

Usage Factor Thermal

Ratchet

Design

Level A / Level B

Level C 2)

Level D

Test



10.3 Reactor Vessel Flange & Stud Bolts

This section describes the stress evaluation results for the RV Flange and Stud Bolts. The stud bolt size is . Figure 10-3-1 shows the RV dimensions and Figure 10-3-2 shows the dimensions of RV flanges.



Figure 10-3-1 Reactor Vessel Dimensions

Dimensions: inch



Figure 10-3-2 Reactor Vessel Flange & Stud Bolts Dimensions



Figure 10-3-3 Reactor Vessel Flange & Stud Bolt Finite Element Model



Figure 10-3-4 Reactor Vessel Flange & Stud Bolt Finite Element Model



Figure 10-3-5 Finite Element Model for RV Flange & Stud Bolt




The calculated stress divided by allowable value results (Stress-to-Allowable Ratio) for the most limiting locations in the Reactor Vessel Flanges are summarized in Table 10-3-1 and Table 10-3-2.

Table 10-3-1 Summary of Results for Reactor Vessel Flanges


Primary Stress Primary plus Secondary

Stress

Fatigue

Condition Part

Pm PL or PL+Pb

Triaxial Stress


Bearing Stress

Shear Stress

Thermal Ratchet

RV Flange Design

RVCH Flange

RV Flange Level A & B

RVCH Flange

RV Flange Level C

RVCH Flange

RV Flange Level D

RVCH Flange

RV Flange Test

RVCH Flange



Table 10-3-2 Summary of Results for Stud Bolts

Condition Category Actual Value

-to-Allowable Ratio

Design Bolt cross sectional area

Average Service Stress

Max Service Stress Level A & B

Fatigue Usage Factor

Average Service Stress Level C

Max Service Stress

Average Tensile Stress

Max Tensile Stress

Average bolt shear

Combined tensile and shear

Distance from bolt center to edge

Level D

Nominal bearing stress



11.0 FRACTURE MECHANICS ASSESSMENT

When the bulk fluid temperature drops to a low level, brittle failure become a concern. Events that are analyzed include heatup / cooldown and the primary & secondary leak tests. For such cases, non-ductile failure evaluations are carried out for postulated flaws of several sizes. The analysis is normally based on a flaw size equal to 1/4 the thickness of the component. In the portions where the flaw size is overly conservative, a smaller flaw size is evaluated. These smaller flaws are selected with consideration of the current inspection techniques. This approach is in accordance with the rules of ASME Section III Appendix-G.



12.0 REFERENCES

1. ASME Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, 2001 Edition through 2003 Addenda.

2. ASME Boiler and Pressure Vessel Code, Section II Material Specification, 2001 Edition through 2003 Addenda.

3. N0-FB10L01 Rev.3, “US-APWR Standard Design General ASME Design Specification for Class 1 and CS Components”.

4. N0-F100L01, Rev.3, “US-APWR Reactor Vessel Design Specification”.

5. SIMULIA, “ABAQUS Analysis User’s manual”, Version 6.7, 2007.

6. Warren M. Rohsenow, James P. Hartnett and Yung I. Cho, “Handbook of Heat Transfer”, Third Edition, McGraw Hill, 1998.

7. (a missing number)

8. An American National Standard, Metric Screw Threads - M Profile, ASME B1.13M-1983.

9. N0-F100A00, Rev.4, “US-APWR Reactor Vessel Outline and Design Drawings”.

muap-09005-np, rev. 1, 'summary of stress analysis results

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